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Konishi, Kensuke; Kunogi, Kosuke
JAEA-Evaluation 2022-005, 106 Pages, 2022/11
JAEA-Evaluation-2022-005.pdf:3.06MB
JAEA-Evaluation-2022-005-appendix(CD-ROM).zip:30.93MB
Japan Atomic Energy Agency (hereafter referred to as "JAEA") consulted with the "Evaluation Committee of Research and Development Activities for Fast Reactor and Fuel Cycle" (hereinafter referred to as "Committee"), which consists of specialists in the fields of the evaluation subjects of fast reactor cycle technologies, for post-review and pre-review assessment of Research and Development (R&D) activities of fast reactor cycle in accordance with "General Guideline for the Evaluation of Government R&D Activities" by Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and Regulation on Conduct for Evaluation of R&D Activities" by JAEA. In response to JAEA's request, the Committee assessed mainly the progress of the R&D project according to guidelines, which addressed the rationale behind the R&D project, the relevance of the project outcome and the efficiency of the project implementation during the period of the current and next plan. This report is issued for the purpose of actively disseminate evaluation information to the people of Japan (based on General Guideline), which lists the members of the Committee and outlines the assessment items and the review process for procedure of the assessment. The assessment reports which were issued by the Committee is attached.
Miyahara, Shinya; Seino, Hiroshi; Ohno, Shuji; Konishi, Kensuke
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
A CONTAIN-LMR code has been developed in JAEA for application to PRA of LMFRs since the original CONTAIN code had been introduced from SNL of U.S. in 1982. The code is a best-estimate, integrated analysis tool for predicting the physical, chemical and radiological conditions inside a containment building of LMFRs following a severe accident with reactor vessel melt-through. The code is also able to predict the source term to the environment in the accident. This code can treat many important phenomena consistently such as sodium fire, radioactive aerosol behavior, hydrogen burn, sodium-concrete reaction and core debris-concrete interaction occurred in the accident with inter-cell heat and mass flow under the multiple cell geometry. This paper describes the chronology of the code development in JAEA briefly as an introduction, and after that, the outline of computational models in the code, the examples of the code validation, and the future plan of the code application are described.
Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Suzuki, Toru; Tobita, Yoshiharu; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; et al.
Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 8 Pages, 2014/12
Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Zuyev, V. A.*; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; Gaidaichuk, V. A.*; et al.
Journal of Nuclear Science and Technology, 51(9), p.1114 - 1124, 2014/09
Times Cited Count：12 Percentile：67.4(Nuclear Science & Technology)Kamiyama, Kenji; Saito, Masaki*; Matsuba, Kenichi; Isozaki, Mikio; Sato, Ikken; Konishi, Kensuke; Zuyev, V. A.*; Kolodeshnikov, A. A.*; Vassiliev, Y. S.*
Journal of Nuclear Science and Technology, 50(6), p.629 - 644, 2013/06
Times Cited Count：19 Percentile：81.37(Nuclear Science & Technology)In core disruptive accidents of sodium cooled fast reactors, fuel discharge from the core region reduces the possibility of severe re-criticality events. In-core coolant channels such as the control-rod guide tube and a concept of the FAIDUS (Fuel Assembly with Inner Duct Structure) provide effective fuel discharge paths if effects of sodium in these paths on molten fuel discharge are limited. Two series of experiments conducted in the present study showed that the discharge path can be entirely voided by the vaporization of a part of the coolant at the initial melt discharge phase, that this is followed by coolant vapor expansion, and that melt penetrates significantly into the voided channel. In conclusion, the effects of the sodium on fuel discharge are limited and therefore in-core coolant channels provide effective fuel discharge paths for reducing neutronic activity.
Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Toyooka, Junichi; Matsuba, Kenichi; Zuyev, V. A.*; Pakhnits, A. V.*; Vurim, A. D.*; Gaidaichuk, V. A.*; Kolodeshnikov, A. A.*; et al.
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12
In order to eliminate energetics potential in the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors, introduction of a fuel subassembly with an inner duct structure has been considered. Recently, a design option which leads molten fuel to discharge upward is considered to minimize developmental efforts for the fuel subassembly fabrication. In this paper, a series of out-of-pile tests and one in-pile test were presented. The out-of-pile tests were conducted to investigate the effects of driving pressures on upward discharge, and the in-pile test was conducted to demonstrate a sequence of upward discharge behavior of molten-fuel. Based on these experimental results, it is concluded that the most of molten-fuel is expected to complete discharging upward before core melting progression, and thereby, introduction of the fuel subassembly with the upward discharge duct has the sufficient potential to eliminate energetics events.
Matsuba, Kenichi; Kamiyama, Kenji; Konishi, Kensuke; Toyooka, Junichi; Sato, Ikken; Zuev, V. A.*; Kolodeshnikov, A. A.*; Vasilyev, Y. S.*
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 7 Pages, 2012/12
A series of fragmentation tests (FR tests) for molten oxide was conducted to obtain experimental knowledge on the distance for fragmentation of molten core material discharged into the lower sodium plenum. Approx. 7
14 kg of molten alumina was discharged into a sodium pool (depth: 1.3 m, diameter: 0.4 m, temperature: approx. 673 K) through a duct (inner diameter: 4063 mm). The test results showed that the molten alumina was fragmented into particles within 1 m from the surface of the sodium pool. The estimated distances for fragmentation in the FR tests were roughly 6070% lower than the predictions by the existing representative correlation. The experimental knowledge confirms the possibility that the distance for fragmentation of the molten core material can be significantly reduced due to thermal interactions in the lower sodium plenum.
Sato, Ikken; Tobita, Yoshiharu; Konishi, Kensuke; Kamiyama, Kenji; Toyooka, Junichi; Nakai, Ryodai; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vassiliev, Y. S.*; et al.
Journal of Nuclear Science and Technology, 48(4), p.556 - 566, 2011/03
In the JSFR design, elimination of severe recriticality events in the Core Disruptive Accident (CDA) is intended as an effective measure to assure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the Initiating Phase selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, with introduction of Inner Duct on the other hand. The effectiveness of these measures are reviewed based on existing experimental data and evaluations performed with validated analysis tools. It is judged that the present JSFR design can exlude severe power burst events.
Toyooka, Junichi; Kamiyama, Kenji; Konishi, Kensuke*; Tobita, Yoshiharu; Sato, Ikken
Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM), 7 Pages, 2010/11
In this study, the heat flux from the molten core materials to the outer surface of the inner duct (the pool-to-duct heat flux) was evaluated based on all the EAGLE-1 in-pile experiments available. Through the evaluation, it was understood that the pool-to-duct heat flux was so high in all the in-pile experiments that the duct wall failed without coolant boiling in its behind. It was also indicated that the presence of steel in the molten core mixture played a key role in this high heat flux. Application of the SIMMER-III code for these tests suggested that some model improvements were necessary to simulate pool-to-duct heat transfer behavior in a mechanistic way.
Kamiyama, Kenji; Isozaki, Mikio; Imahori, Shinji; Konishi, Kensuke; Matsuba, Kenichi; Sato, Ikken
JAEA-Research 2008-059, 33 Pages, 2008/07
JAEA-Research-2008-059.pdf:10.82MB
In CDA of LMFBR, molten core materials would discharge from the core region through the coolant paths. Rapid vaporization of the coolant by mixing of the molten core materials provides effective evacuation of the liquid coolant from the path and reduces significantly possibility of core-material freezing and blockage formation inside the paths. This characteristic enhances early discharge of molten-core materials and reduces possibility of severe re-criticality events. In this study, melt discharge experiments were conducted with a coolant channel simulating the discharge path with an enhanced length of the path compared with that of the realistic design structure. An alloy and water were used as simulant of the molten fuel and sodium respectively. This series of experiments showed that the discharge path can be entirely voided by vaporization of a part of the coolant at the initial melt discharge phase, followed by vapor expansion toward the end of the coolant channel. Furthermore, it was revealed that the condition where coolant void expansion started can be defined by melt-coolant sensible heats ratio and the heated height of the coolant. The heat balance evaluation during the coolant void expansion phase shows that the film condensation heat transfer should be considered. The coolant-void-expansion behavior in the discharge path of the realistic design condition was estimated based on an application of this knowledge to existing experiments with molten oxide and sodium.
Konishi, Kensuke; Toyooka, Junichi; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.
Nuclear Engineering and Design, 237(22), p.2165 - 2174, 2007/11
Times Cited Count：42 Percentile：92.58(Nuclear Science & Technology)The WF (Wall Failure) test of the EAGLE program, in which 2kg of uranium dioxide fuel-pins were melted by nuclear heating, was successfully conducted in the IGR of NNC/Kazakhstan. In this test, a 3mm-thick stainless steel (SS) wall structure was placed between fuel pins and a 10mm-thick sodium-filled channel (sodium gap). During the transient, fuel pins were heated, which led to the formation of a fuel-steel mixture pool. Under the transient nuclear heating condition, the SS wall was strongly heated by the molten pool, leading to wall failure. The time needed for fuel penetration into the sodium-filled gap was very short (less than 1 second after the pool formation). The result suggests that molten core materials formed in hypothetical LMFBR core disruptive accidents have a certain potential to destroy SS-wall boundaries early in the accident phase, thereby providing fuel escape paths from the core region. The early establishment of such fuel escape paths is regarded as a favorable characteristic in eliminating the possibility of severe re-criticality events.
Konishi, Kensuke; Kubo, Shigenobu*; Sato, Ikken; Koyama, Kazuya*; Toyooka, Junichi; Kamiyama, Kenji; Kotake, Shoji*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.
Proceedings of 5th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-5), p.465 - 471, 2006/11
no abstracts in English
Konishi, Kensuke; Toyooka, Junichi; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.
Proceedings of Technical Meeting on Severe Accident and Accident Management (CD-ROM), 16 Pages, 2006/03
no abstracts in English
Morita, Koji*; Matsumoto, Tatsuya*; Fukuda, Kenji*; Tobita, Yoshiharu; Yamano, Hidemasa; Konishi, Kensuke; Sato, Ikkenn
JNC TY9400 2003-011, 56 Pages, 2003/04
JNC-TY9400-2003-011.pdf:2.31MB
It is one of the important problems for more reliable reactor safety evaluation to improve numerical simulation techniques for involved thermal-hydraulic phenomena of multiphase, multicomponent flows in core disruptive accidents.In the present joint research, physical model development and experimental investigation were conducted for transient condensation phenomena of a vapor bubble with noncondensable gases to improve applicability of a fast-reactor safety analysis code for the phase-transition phenomena in multicomponent systems.In this fiscal year, preliminary experiments using noncondensable gas were performed for the transient bubble condensation phenomena, and then basic data were obtained for large-scale bubble behavior without condensation.In addition, a multiple-scale flow-regime model treating large-scale bubbles was newly proposed for the fast-reactor safety analysis code and applied to the analysis of the preliminary experiments successfully.
; Konishi, Kensuke
JNC TN9400 2001-035, 65 Pages, 2001/02
JNC-TN9400-2001-035.pdf:2.64MB
Fast neutron hodoscope system as the main fuel motion diagnostic system in the Safety Engineering Reactor for Accident Phenomenology (SERAPH) was studied in order to realize high-resolution diagnostics for fuel density variation according to fuel motion. Based on the study on minimization of the detector size and improvement for the detector sensitivity, optimized concept of the detector design was proposed. 0n the other hand, geometry of the collimator and slot system was also optimized through the evaluation of optical conditions. Using the optimized design of the detector, collimator, and slot system, diagnostic performance for main SERAPH tests (single pin test, pin bundle test and molten fuel pool test, etc.) were quantified. It was confirmed that proposed system has sufficient diagnostic capability required in the SERAPH tests.
Mishima, Kaichiro*; Hibiki, Takashi*; Saito, Yasushi*; Tobita, Yoshiharu; Konishi, Kensuke; Suzuki, Toru
JNC TY9400 2000-018, 72 Pages, 2000/07
JNC-TY9400-2000-018.pdf:3.92MB
no abstracts in English
Konishi, Kensuke; Toyooka, Junichi; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Pakhnits, A. V.*; Gaidaichuk, V. A.*; et al.
no journal, ,
no abstracts in English
Kamiyama, Kenji; Konishi, Kensuke; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Shimakawa, Yoshio*; Koyama, Kazuya*; Zuyev, V.*; Vassiliev, Y. S.*; Kolodeshnikov, A.*
no journal, ,
no abstracts in English
Toyooka, Junichi; Konishi, Kensuke; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Pakhnits, A. V.*; Gaidaichuk, V. A.*; et al.
no journal, ,
no abstracts in English
Konishi, Kensuke; Toyooka, Junichi; Kamiyama, Kenji; Sato, Ikken; Kubo, Shigenobu*; Kotake, Shoji*; Koyama, Kazuya*; Vurim, A. D.*; Gaidaichuk, V. A.*; Pakhnits, A. V.*; et al.
no journal, ,
no abstracts in English
